ML20211D544

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Rev 1 to, Conformance to Reg Guide 1.97,JM Farley Nuclear Plant,Units 1 & 2, Technical Evaluation Rept
ML20211D544
Person / Time
Site: Farley  Southern Nuclear icon.png
Issue date: 11/30/1986
From: Stoffel J
EG&G IDAHO, INC.
To:
NRC
Shared Package
ML20211D513 List:
References
CON-FIN-A-6483, RTR-REGGD-01.097, RTR-REGGD-1.097 EGG-EA-6794, EGG-EA-6794-R01, EGG-EA-6794-R1, TAC-51088, TAC-51089, NUDOCS 8702240024
Download: ML20211D544 (50)
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EGG-EA-6794

  • TECHNICAL EVALUATION REPORT CONFORMANCE TO REGULATORY GUIDE 1.97 JOSEPH M. FARLEY HUCLEAR PLANT, UNIT NOS. 1 AND 2 Docket Nos. 50-348/364 J. W. Stoffel Published November 1986 Idaho National Engineering Laboratory EG&G Idaho, Inc.

Idaho Falls, Idaho 83415 Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Under DOE Contract No. DE-AC07-761001570 FIN No. A6483 ADOhKOhhoh4g PDR

a ABSTRACT This EG&G Idaho, Inc., report reviews the submittals for Regulatory Guide 1.97 for Unit Nos. I and 2 of the Joseph M. Farley Nuclear Plant and identifies areas of nonconformance to the regulatory guide. Exceptions to Regulatory Guide 1.97 are evaluated and those areas where sufficient basis for acceptability is not provided are identified.

1 Occket Nos. 50-348 and 50-364 TAC Nos. 51088 and 51089 l l

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FOREWORD

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This report is supplied as part of the " Program for Evaluating Licensee / Applicant Conformance to R.G. 1.97," being conducted for the U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Division of PWR Licensing-A, by EG&G Idaho, Inc., NRR and,I&E Support Branch. g i

The U.S. Nuclear Regulatory Commission funded the work under authorization 20-19-10-11-3. .

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Docket No. 50-348 and 50-364 TAC No. 51088 and 51089 i

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CONTENTS ABSTRACT .............................................................. 11 FOREWORD .............................................................. iii

1. INTRODUCTION ..................................................... 1
2. REVIEW REQUIREMENTS .............................................. 2

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3. EVALUATION ....................................................... 4 3.1 Adherence to Regulatory Guide 1.97 ........................ 4 3.2 Type A Variables ........................................... 4 -

3.3 Exceptions to Regulatory Guide 1.97 ........................ 5

4. CONCLUSIONS ...................................................... 19
5. REFERENCES ....................................................... 20 L

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CONFORMANCE'TO'REGULAT_0RY GUIDE l'.97 JOSEPH M. FARLEY\ NUCLEAR PLANT UNIT NOS 1 AND 2 7 ,

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4 i-m 1. INTRODUCTION ,?

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s -- On Deceder-17, 1982, Generic Letter No. 82-33 (Reference 1) was s

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. , issued by 0, C. Eisenhut, Director.of the Division of Licensing, Nuclear '

Rear. tor Regulation, to all licenseat of operating reactors, applicants for

}c'pyratinglicens'esandholdersofconstructionpermits. This letter ,

%1ncluded: additional clarification regarding Rdgulatory Guide 1.97, W Revision'2 (Reference 2), relating to the requirements for emergency

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response capability. These requirements have been published as Supplement lh No. 1 to NUREG-0737, "TMI Action Plan Requirements" (Reference 3).

AlabamaPowerCompany,the~ lice $sicfortheJosephM.FarleyNuclear Plant, Unit Nos. 1 and 2, provided responses to't'he Regulatcry Guide 1.97 portion of.the ge6eric letter for Unit No. 2'on June 29, 1984 (Reference 4) and for Unit No.i on March 30, 1984 (Reference 5). Additior.a1 information was provided on April 10, 1985 (Reference 6) and Augsit 8, 1986 (Reference 7). s F t This report provides an evaluattun of these submittals.

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2. Rt' VIEW REQUIREMENTS Section 6.2 of NUREG-0737, Supplement No. 1, sets forth the documentation to be submitted in a report to the NRC describing how the licensee complies to Regulatory Guide 1.97 as applied to emergency response facilities. The submittal should include documentation that provides the ,

following information for each variable shown in the applicable table of Regulatory Guide 1.97:

1. Instrument range
2. Environmental qualification
3. Seismic qualification I
4. Quality assurance
5. Redundance and sensor location
6. Power supply
7. Location of display
8. Schedule of installation or upgrade.

Furthermore, the submittal should identify deviations from Regulatory Guide 1.97 and provide supporting justification or alternatives.

Subsequent to the issuance of the generic letter, the NRC held regional meetings in February and March 1983, to answer licensee and applicant questions and concerns regarding the NRC policy on this subject. -

At these meetings, it was noted that the NRC review would only address exceptions taken to Regulatory Guide 1.97. Furthermore, where licensees or applicants explicitly state that instrument systems conform to the regulatory guide it was noted that no further staff review would be necessary. Therefore, this report only addresses exceptions to Regulatory

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e Guide 1.97. The following evaluation is an audit of the licensee's submittals based on the review policy described in the NRC regional meetings.

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3. EVALUATION The licensee provided responses to NRC Generic Letter 82-33 on

\ June 29, 1984 (Unit 1), March 30, 1984 (Unit 2), April 10, 1985 and August 8, 1986 (Units 1 and 2). This evaluation is based on these submittals.

3.1 Adherence to Reaulatory Guide 1.97 The licensee stated that compliance with Regulatory Guide 1.97 is indicated on their review checklist which summarizes each variable's compliance with the Regulatory Guide 1.97 provisions. That compliance report presents justification, modifications or ongoing evaluations that are provided as resolutions for any identified deviations. Therefore, it {

is concluded that the licensee has provided an explicit commitment on conformance to Regulatory Guide 1.97. Exceptions to and deviations from the regulatory guide are noted in Section 3.3. -

3.2 Tvoe A variables Regulatory Guide 1.97 does not specifically identify Type A variables, i.e., those variables that provide information required to permit the control room operator to take specific manually controlled safety actions.

The licensee classifies the following instrumentation as Type A:

1. Reactor coolant system (RCS) pressure (wide range)
2. RCS hot leg temperature (wide range)
3. RCS cold leg temperature (wide range)
4. Steam generator level (wide range) .
5. Steam generator level (narrow range) 4

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6. Pressurizer. level
7. Containment pressure (nornal range) l l
8. Main steamline pressure
9. Refueling water storage tank level l

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10. Containment water level
11. Condensate storage tank level
12. Auxiliary feedwater flow
13. Core exit temperature
14. Core subcooling monitor The above instrumentation meets Category 1 requirements consistent with the requirements for Type A variables, except as noted in Section 3.3.

3.3 ExceDtions to Reaulatory Guide 1.97 4

The licensee identified deviations and exceptions from Regulatory Guide 1.97. These are discussed in the following paragraphs.

3.3.1 Environmental Qualification Reautrement Deviation In References 4 and 5, the licensee has indicated that environmental qualification is not applicable for the following Category 1 and 2

. instrumentation. However, no justification was submitted for this 4 deviation. In Reference 7, the licensee stated that the instrumentation listed, along with the associated instrument loop components, are located outside areas that constitute a harsh environment.

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Main steamline pressure Refueling water storage tank level Condensate storage tank level Plant vent stack flow Condenser steam jet air ejector radiation ~

Plant vent effluent radiation .

Accessible area radiation Main steam effluent radiation Turbine driven auxiliary feedwater effluent radiation Heating, ventilating and air conditioning emergency damper position--control room Pressurizer heater breaker position Status of standby power and other energy sources important to safety.

Based on the licensee's justification that the instrumentation listed is located in a mild environment, we find this deviation acceptable.

Deviations other than environmental qualification for these variables are listed elsewhere in this report.

3.3.2 Neutron Flux (Intermediate Ranae)

The installed neutron flux instrumentation does not completely meet the redundancy requirements of Regulatory Guide 1.97. Both intermediate range instrument loops are ultimately powered from the same DC power supply train (Train A). The power to the instrument loops is provided by separate inverters, and the outputs of these inverters are physically separated and backed up by diesel generator A. In addition, an alternate source of power, other than the inverters, is provided to both instrument loops from a Solatron voltage regulator. The licensee is installing a third channel of wide range instrumentation to resolve ambiguity between the existing instrumentation should one loop fail. This new instrumentation loop,

  • however, will be powered from the same power supply train (Train A) as the two existing neutron flux monitoring loops. The licensee also states that the existing electrical independence of the neutron monitors is consistent with the design criteria of the reactor protection system.

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l There is a very low probability of a fault occurring that would fail the complete electrical train A (ac and dc). In the event this loss should occur, only the intermediate range of neutron flux instrumentation would be lost. No single failures such as inverters, batteries, battery chargers, 4160 volt bus faults or dc bus faults would disable all the neutron flux

instrumentation. Based on this evaluation, we conclude that the existing configuration of this variable is acceptable.

3.3.4 RCS Soluble Boron Concentration Regulatory Guide 1.97 recommends Category 3 instrumentation, with a range of 0 to 6000 ppm, for this variable. The licensee takes credit for the post-accident sampling system to meet this recommendation.

The licensee takes exception to Regulatory Guide 1.97 with respect to post-accident sampling capability. This exception goes beyond the scope of this review and is being addressed by the NRC as part of their review of NUREG-0737, Item II.B.3.

3.3.5 RCS Cold and Hot Leo Water Temperature The maximum indication of the instrumentation for these variables is i 700*F. This is 50*F less than the Regulatory Guide 1.97, Revision 2, range j

guidelines (50 to 750*f).

l I Regulatory Guide 1.97, Revision 3. (Reference 8) recommends a range of 50 to 700*F for these variables. The instrumentation supplied by the licenses meets this recommended range and is, therefore, acceptable.

l 3.3.6 Coolant Level in Reactor (Unit No. 2)

The licensee does not have usable instrumentation for this variable.

As justification, the licensee states that they participated in a pilot project for a non-invasive reactor vessel level system. This unsuccessful demonstration led the licensee into a detailed review of commercially available reactor vessel level systems. The results of this ongoing review 7

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indicate to the licensee that no commercially available reactor vessel level system has been accepted by the NRC for operational use.

It is our understanding that two systems are now commercially available for reading reactor vessel level in a pressurized water reactor.

One system uses heated junction thermocouples (NUREG/CR-2627, Reference 9) -

and the other system uses differential pressure (NUREG/CR-2628, Reference 10).

The NRC is reviewing the acceptability of this variable as part of

' their review of NUREG-0737, Item II.F.2.

3.3.7 Decrees of Subcoolino The licensee has identified degrees of subcooling as a Type A variable. As such, it should meet Category 1 requirements. The licensee states that their core subcooling monitor meets Category 2 requirements.

The NRC is reviewing the acceptability of this variable as part of their review of NUREG-0737, Item II.F.2.

3.3.8 Containment Sumo Water Level The licensee has taken exception to the range recommended by Regulatory Guide 1.97 for the containment level instrumentation (bottom of containment to 600,000 gallon level equivalent). The licensee has instrumentation with a minimum level indication of 62.000 gallons. The licensee considers the existing range to be adequate since the minimum level indication is limited by physical installation restraints of the float type level measurement device and no operator actions are required below the 62,000 gallon level.

The reactor cavity sump level indication would provide a diverse method of determining a water level increase in the containment. Since no 8

operator action is required at less than the minimum indication available with the existing range, we find this to be an acceptable deviation from Regulatory Guide 1.97.

3.3.9 Containment Isolation Valve Position The licensee has not provided redundancy for all of the containment

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  • isolation valves. Some isolation valves ",nside and outside containment for the same penetration are of the same train orientation and, therefore, redundant indication is not provided. The licensee submitted the following ju".tification for this deviation. These valves are normally closed valves and remain closed in an accident condition until remotely opened by the cperator. The power supply for these valves is for position indication as well as for power operation of the valve motor operators. The valves are part of a penetration which is redundant to another penetration. At least cne of these redundant piping systems must be opened during certain accident conditions. Therefore, the power for both containment isolation valves on a penetration is from the same power supply to ensure that a single power supply failure will not inhibit both penetrations from operating. Both isolation valves for the redundant penetration are supplied power from another power source.

We find the licensee's justification acceptable. Furthermore, if during an accident condition, a single train of electrical power were to fail resulting in a loss of position indication, the operator could verify that the outside containment isolation valve is closed and containment integrity maintained. Therefore, this is an acceptable deviation from Regulatory Guide 1.97.

I 3.3.10 Radioactivity Concentration or Radiation Level in Circulatina Primary Coolant

. The licensee uses the post-accident sample system to measure this parame ter . In a letter dated February 17, 1984, the licensee states that procedures exist which relate radionuclide concentrations to core damage.

These procedures consider physical parameters such as core temperature and 9

sample locations. Alabama Power Company will implement a calculational method to assess the extent of core damage. This method will utilize the RCS post-accident sampling system in the determination of the status of fuel cladding and the magnitude of any core damage.

Based on the alternate instrumentation provided by the licensee, we

. conclude that the instrumentation supplied for this variable is adequate and, therefore, acceptable.

3.3.11 Containment Hydrogen Concentration Regulatory Guide 1.97 recommends Category 1 instrumentation for this variable with a range of 0 to 10 percent. The licensee has installed Category 3 hydrogen analyzers that do not meet the range recommended by 3 Regulatory Guide 1.97. The licensee considers this instrumentation j acceptable because the operators energize the hydrogen recombiners based on loss of coolant accident (LOCA) indications. Hydrogen concentration is not a LOCA indication, it is used only as the basis for verifying the hydrogen .

removal capability of the hydrogen recombiners. In the event that the hydrogen analyzers are unavailable to provide containment hydrogen concentration, sufficient time is available to determine the containment hydrogen concentration utilizing the containment air post-accident sampling system (CAPASS).

The NRC has reviewed t.se acceptability of this variable as part of their review of NUREG-0737. Item II.F.1.6.

3.3.12 Residual Heat Removal (RHR) Heat Exchanaer Outlet Temperuture Regulatory Guide 1.97 Revision 2, recommends a range for this variable of 32 to 350*F. Revision 3 changed the recommended range to 40 to 350*F. The licensee has supplied a range of 50 to 400*F. The instrumentation supplied has a range where the lower limit of the span does not conform to either revision of the regulatory guide. The licensee 10

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, states that the existing range of this instrumentation envelops the RHR system design parameters and no need exists for temperature indication below 50*F.

Based on the justification provided by the licensee, we conclude that  ;

the instrumentation supplied for this variable is adequate to monitor this l variable during all accident and post-accident conditions, and is

,. therefore, acceptable.

3.3.13 Accumulator Tank !.evel and Pressure 4

The licensee has Category 3 accumulator tank level instrumentation

, that does not meet the recommended range. The justification submitted by the licensee for this deviation is that the accumulator tank level at the Farley Nuclear Plant was designed solely to verify compliance with the technical specification volume provisions. In the event of RCS depressurization, accumulator tank discharge is verified by monitoring accumulator tank pressure, which meets the Category 2 requirements. -

The accumulators are passive and discharge for RCS breaks. The level l and pressure measurement channels are not required to protect the integrity of the RCS boundary, to shutdown the reactor, to maintain it in a safe shutdown condition or to prevent or mitigate the consequences of an accident which could result in potential exposures. We find the qualified pressure instrumentation supplied for this variable adequate to determine i that the accumulators have discharged. Therefore, the existing i instrumentation is acceptable to monitor this variable.

3.3.14 Refuelina Water Storace Tank Level Regulatory Guide 1.97 recommends a range from the top to the bottom of the tank. The licensee does not meet this range and states that the

. maximum level indication of the existing instrumentation is one foot below l

the top of the tank. This level indication reads from 0 to 40 feet and envelopes the technical specification volume requirement, which the licensee states is sufficient to mitigate any design basis event. This 11

range is adequate to provide the operator with information for normal operations and to perform switchover from emergency core cooling system injection to recirculation.

Based on the licensee's justification, we conclude that the existing instrumentation for this variable (that reads 98 percent of the recommended .

. range) is adequate to monitor this variable during all accident and post-accident conditions.

5.3.15 Pressurizer Level Regulatory Guide 1.97 recommends a range from the bottom to the top for this variable. The instrumentation provided by the licensee does not read this full range. The licensee states that the volume measured repiesents approximately 89 percent of the pressurizer and is sufficient for the operator to take the required manual actions and to ensure the proper operation of the pressurizer.

The portion of the pressurizer level that is not indicated (approximately 11 percent) is the upper and lower hemispherical head region, where the volume to level ratio is not linear. We find this deviation minor and acceptable. The existing range is adequate to monitor this variable during all accident and post-accident conditions.

3.3.16 Pressurizer Heater Status Regulatory Guide 1.97 recommends electric current instrumentation to determine the operating status of the pressurizer heaters. The licensee does not intend to provide specific instrumentation to read this current.

The licensee states, in Reference 4 and 5, that the pressurizer heater status can be adequately determined by the use of pressurizer heater circuit breaker position and pressurizer pressure. Furthermore, the licensee states that the emergency response procedures do not utilize pressurizer heater current for accident mitigation.

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,a In Reference 7, the licensee lists several additional means of determining pressurizer heater operation. In addition to these, the licensee states that heaters current can be monitored with the diesel generator megawatt and current indicators when the diesel generator is supplying power for the heaters and with the 4.16 KV bus incoming current indicator when being supplied from offsite power.

, Based on the available current monitoring instrumentation and the diverse reans of determining heater operation, we conclude that this is an acceptable deviation from Regulatory Guide 1.97.

3.3.17 Ouench Tank Level The range of the existing instrumentation for this variable does not meet the range recommended by Regulatory Guide 1.97 (top to bottom). The licensee's justification for this deviation is that only 5 percent of the total tank volume is not measured and the existing range is sufficient to provide the operator with the necessary information for accident monitoring.

We find the existing level range adequate to monitor the operation of this tank. Therefore, this is an acceptable deviation from Regulatory Guide 1.97.

3.3.18 Steam Generator Level Regulatory Guide 1.97 recommends a range from the tube sheet to the separators for this variable. The licensee has instrumentation that reads from 12 in, above the tube sheet to the separators. The licensee states that the volume of the steam generator not measured is less than 2 percent of the volume recommended by the regulatory guide.

The steam generator is, in effect, empty at 12 in, above the tube

. sheet; therefore, this deviation is minor with respect to the overall range and system accuracy. The existing range is adequate to monitor this variable during all accident and post-accident conditions.

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3.3.19 Steam Generator Pressure The licensee has instrumentation for this variable with a range of 0 to 1200 psig. Regulatory Guide 1.97 recommends a range from atmospheric pressure to 20 percent above the lowest safety valve setting. Thus, the range should be to 1290 psig, as the licensee identifies the range as being .

,90 psig less than recommended. The licensee justifies the range deviation by stating that the highest actuation setpoint of the main steam safety valves is 1129 psig. Allowing for 3 percent accumulation the auximum credible steamline pressure is 1163 psig which is within the indicated range of the existing instrumentation.

, Based on the licensee's statement that the maximum credible steamline pressure would be 1163 psig, this instrumentation would remain on scale q during any accident or post-accident conditions. Therefore, we find this deviation acceptable.

3.3.20 Volume Control Tank Level The licensee tskes exception to the range recommended by Regulatory I Guide 1.97 for this variable (top to bottom). The transmitters measure the full range between the instrument connections, however. these connections are not at the top and bottom of the tank. The justification submitted by the licensee is that for operational purposes, level indication at either end of the scale is considered full or empty. Also, the existing range of the volume control tank level envelops all automatic action of the level control system.

We find that the existing level indication is adequate to monitor the operation of this tank. Therefore, this is an acceptable deviation from Regulatory Guide 1.97.

3.3.21 Hiah Level Radioactive Liauid Tank Level Regulatory Guide 1.97 recommends a range of top to bottom for this variable. The transmitters for these tanks measure the full range between 14

> l the instrument connections; however, these connections are not at the top and bottom of the tank. The licensee's justification for this deviation is that at-least 90 percent of the tank volume is measured and the range is sufficient to provide the operator with the necessary information for accident monitoring.

4 We find that the existing range is adequate to monitor the operation

,cf this tank during all accident and post-accident conditions. Therefore, this is an acceptable deviation from Regulatory Guide 1.97.

3.3.?? Padioactive Gas Holdup Tank pressure I

The licensee takes exception to the range recommended by Regulatory Guide 1.97 for this variable (0 to 150 percent design pressure). The I licensee has instrumentation for this variable that reads from 0 to 100 percent of the destga pressure of the tank (150 psig). The licensee states that the existing range is acceptable because it covers up to the design pressure of the tanks and because relief valves are installed on cach tank to prevent the tank pressure from exceeding the design value of 150 psig.

Based on the justification provided by the licensee, we conclude that

the instrumentation provided for this variabic is adequate to monitor the cperation of this tank and is, therefore, acceptable.

3.3.23 Radiation Exposure Rate

! In References 4.and 5, the licensee identified two deviations for this variable. First, of.the plant areas which are accessible post-accident, cnly the control room has a permanently installed radiation monitor.

Second, the range of the radiation level indication of the control room

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radiation monitor is 10

  • to 10 R/hr. The range specified by

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Regulatory Guide 1.97 for this variable is 10"I to 10* R/hr.

In Reference 7, the Itcensee identified permanently installed monitors in areas required for post-accident access. The ranges of these i

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l instruments are less than recommended by the regulatory guide, however the licensee states that the ranges are adequate since additional shielding has been ir, stalled in these areas and portable monitoring instruments would be used to assess radiation levels before entry into these areas.

The licensee has shown an analysis of radiation levels expected for the monitor locations. The existing radiation exposure rate monitors have

' ranges that encompasses the expected radiation levels in their location.

Based on this, and the fact that personnel would not be permitted into the area without portable monitoring if the upper limit of the range is exceeded, we find the instrumentation provided for this variable acceptable.

3.3.24 Plant and Environs Radiation (Portable Instrumentation) 4 Regulatory Guide 1.97 recommends a range of 10-3 to 10 rads /hr, for beta radiation and low energy photons. The licensee states that the maximum indication of the existing portable instrumentation is below the recommended maximum level. The licensee's justification for this deviation .

is that their portable instrumentation has sufficient range to monitor the radiation levels in areas of the plant where post-accident access is necessary by plant personnel.

This instrumentation is portable and would not be used to assess levels of radiation greater than the range provided by the licensee.

Therefore, this is an acceptable deviation from Regulatory Guide 1.97.

3.3.25 Plant and Environs Radioactivity (Portable Instrumentation)

The licensee does not have a portable multichannel gamma-ray spectrometer, as recommended by Regulatory Guide 1.97, Revision 2, for this variable. Regulatory Guide 1.97, Revision 3, states that portable instrumentation should be provided for isotopic analysis of plant and environs radioactivity. The licensee has also not provided portable ,

instrumentation for isotopic analysis. However, the licensee does have two non-portable multichannel analyzers (MCA) located in the counting room of 16

the plant. The MCAs are equipped with a germanium-lithium detector to provide isotopic analysis of the plant and environ samples. The MCAs have the capability to analyze samples in less than 15 minutes from the time the sample is delivered to the MCAs. The MCAs located in the plant are used 4 during normal plant operations, are accessible post-accident, and are instruments familiar to plant personnel.

The licensee states that a portable multichannel gamma-ray spectrometer would not enhance the capability to perform isotopic analysis. A portable device can only provide " scoping" of the radionuclide content and cannot provide a quantitative measurement. The existing non-portable MCAs at the farley Nuclear plant would provide a quantitative measurement of the radionuclide content.

i The two existing multichannel analyzers are sufficient to provide for isotopic analysis and an adequate and timely assessment of radioactive releases at this station. Therefore, this is an acceptable deviation from Regulatory Guide 1.97. .

3.3.26 Wind Speed l

l Regulatory Guide 1.97, Revision 2, recomends a range of 0 to 30 meters /second (67 mph) for this variable. The licensee has instrumentation with a range of 0 to 22 meters /second (50 mph). The licensee justifies this deviation by stating that their existing wind speeo instrumentation has historically provided reliable indications that are representative of meteorological conditions in the plant vicinity.

Regulatory Guide 1.97, Revision 3, recomends instrumentation with a range of 0 to 22 meters /second (50 mph) for this variable. Since the existing instrumentation meets the Regulatory Guide 1.97, Revision 3 requirement, this deviation is acceptable.

3.3.27 Accident Samplina (Primary Coolant. Containment Air and Sump)

The minimum quantifiable concentrations of boron, chlorides, dissolved hydrogen, total gas and oxygen do not meet Regulatory Guide 1.97 range 17

guidelines. The licensee states that analysis below the minimums identified would serve no useful purpose for accident analysis, mitigation or recovery.

The minimum quantifiable concentrations of oxygen and hydrogen in the containment air do not satisfy Regulatory Guide 1.97 range guidelines. The licensee's justification for this deviation is that the minimum quantifiable concentrations represent the minimum detectable concentrations. In addition, the itcensee states that analysis below the identified minimums would serve no useful purpose for accident analysis, mitigation or recovery, i

The licensee takes exception to Regulatory Guide 1.97 with respect to post-accident sampling capability. These exceptions go beyond the scope of this review and are being addressed by the NRC as part of their review of NUREG-0737, Item II.B.3.

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4. CONCLUSIONS Based on our review, we find that the licensee conforms to, or is justified in deviating from Regulatory Guide 1.97.

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5. REFERENCES
1. NRC letter, D. G. Eisenhut to All Licensees of Operating Reactors, Applicants for Operating Licenses, and Holders of Construction Pern.its, " Supplement No. 1 to NUREG-0737--Requirements for Emergency Response Capability (Generic Letter No. 82-33)," December 17, 1982.
2. Instrumentation for Licht-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions Durina and Followino an Accident,
3. Clarification of TMI Action Plan Reautrements. Reauirements for Emercency Response Capability, NUREG-0737, Supplement No. 1, NRC, Office of Nuclear Reactor Regulation, January 1983.
4. Alabama Power Company letter, R. P. Mcdonald to Director, Nuclear Reactor Regulation, NRC, " Regulatory Guide 1.97 Compliance," June 29, 1984.
5. Alabama Power Company letter, F. L. Clayton, Jr. to Director, Nuclear Reactor Regulation, NRC, " Regulatory Guide 1.97 Compliance,"

March 30, 1984.

6. Alabana Power Company letter, R. P. Mcdonald to Director, Nuclear Reactor Regulation, NRC, " Regulatory Guide 1.97 Compliance," April 10, ,

1985.

7. Alabama Power Company letter, R. P. Mcdonald to Director, Nuclear Reactor Regulation, NRC, " Regulatory Guide 1.97 Compliance".

August 8, 1986.

8. Instrumentation for Licht-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions Durina and Followino an Accident, Regulatory Guide 1.97, Revision 3, NRC, Office of Nuclear Regulatory Research, May 1983.
9. Inadeauate Core Coolino Instrumentation Usina Heated Junction Thermocouples for Reactor Vessel Level Measurement, NUREG/CR-2627, ORNL/TM-8248, March 1982.
10. Inadeauate Core Coolino Instrumentation Usina Differential Pressure for Reactor vessel Level Measurment, NUREG/CR-2628, ORNL/TM-8269, March 1982.

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&*%3' BIBLIOGRAPHIC DATA SHEET Revision 1 tttitstiuCTiCN50amt=tatweest 3117t8 #wM SwStaYkt 3 Llavt 86ANK Confornance to Regulatory Guide 1.97, Joseph M. Farley Nuclear Plant

  • CartalPo " Co- tito Unit Nos. 1 & 2 .o~r. .taa g

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November 1986 J. W. Stoffel .so~ , . ,taa g

November 1986 7 pt.~sca sis d e OmGases2af sO4 8eaut amo esaiki=G AOcatsS nariwee te cesse a PacatCT. Tass /vooms wait wwweta EG&G Idaho, Inc. ,,,,,,,,,,,,,,,,,,,,,

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Division of PWR Licensing A Technical Evaluation Report Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, DC 2055C 13 gyppLgwgasf amv MOTES t3 A457w&C7 dJUD meuve er ame, This EGSG Idaho, Inc. report reviews the submittals for the Joseph fl. Farley Nuclear Plant, Unit Nos.1 & 2, and identifies areas of non-conformance to Regulatory Guide 1.97. Any exceptions to these guidelines are evaluated.

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  • TECHNICAL EVALUATION REPORT CONFORMANCE TO REGULATORY GUIDE 1.97 JOSEPH M. FARLEY NUCLEAR PLANT, UNIT NOS. 1 AND 2 Docket Nos. 50-348/364 J. W. Stoffel Published November 1986 Idaho National Engineering Laboratory EG&G Idaho, Inc.

Idaho Falls, Idaho 83415 Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C. 20555 under DOE Contract No. DE-AC07-76IO01570 FIN No. A6483 fV

ABSTRACT This EG&G Idaho, Inc., report reviews the submittals for Regulatory Guide 1.97 for Unit Nos. 1 and 2 of the Joseph M. Farley Nuclear Plant and identifies areas of nonconformance to the regulatory guide. Exceptions to Regulatory Guide 1.97 are evaluated and those areas where sufficient basis for acceptability is not provided are identified.

Docket Nos. 50-348 and 50-364 TAC Nos. 51088 and 51089 11

l FOREWORD

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This report is supplied as part of the " Program fo'r Evaluating Licensee / Applicant Conformance to R.G. 1.97," being conducted for the ,

U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Division of PWR Licensing-A, by EG&G Idaho, Inc., NRR a_nd I&E Support Branch. ,

t The U.S. Nuclear Regulatory Commission funded the work under authorization 20-19-10-11-3. .

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Docket No. 50-348 and 50-364 TAC No. 51088 and 51089 i

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CONTENTS ABSTRACT .............................................................. 11 FOREWORD .............................................................. iii

1. INTRODUCTION ..................................................... 1
2. REVIEW REQUIREMENTS .............................................. 2
3. EVALUATION ....................................................... 4 3.1 Adherence to Regulatory Guide 1.97 ........................ 4 3.2 Type A Variables ........................................... 4 3.3 Exceptions to Regulatory Guide 1.97 ........................ 5
4. CONCLUSIONS ...................................................... 19
5. REFERENCES ....................................................... 20 i

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CONFORMANCE TO REGULATORY GUIDE 1.97 JOSEPH M. FARLEY NUCLEAR PLANT. UNIT NOS. 1 AND 2

1. INTRODUCTION On December 17, 1982, Generic Letter No. 82-33 (Reference 1) was

,. ' issued by D. G. Eisenhut, Director of the Division of Licensing, Nuclear Reactor Regulation, to all licensees of operating reactors, applicants for cperating licenses and holders of construction permits. This letter included additional clarification regarding Regulatory Guide 1.97, Revision 2 (Reference 2), relating to the requirements for emergency response capability. These requirements have been published as Supplement No. 1 to NUREG-0737, "TMI Action P1'an Requirements" (Reference 3).

Alabana Power Company, the licensee for the Joseph M. farley Nuclear Plant, Unit Nos. I and 2, provided responses to the Regulatory Guide 1.97 portion of the generic letter for Unit No. 2 on June 29, 1984 (Reference 4) y and for Unit No. 1 on March 30, 1984 (Reference 5). Additional information was provided on April 10, 1985 (Reference 6) and August 8, 1986 i (Reference 7).

j This report provides an evaluation of these submittals.

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2. Nt' VIEW REQUIREMENTS l

Section 6.2 of NUREG-0737, Supplement No. 1, sets forth the documentation to be submitted in a report to the NRC describing how the licensee complies to Regulatory Guide 1.97 as applied to emergency response facilities. The submittal should include documentation that provides the ,

following information for each variable shown in the applicable table of l Regulatory Guide 1.97:

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1. Instrument range
2. Environmental qualification
3. Seismic qualification
4. Quality assurance
5. Redundance and sensor location i
6. Power supply
7. Location of display
8. Schedule of installation or upgrade.

Furthermore, the submittal should identify deviations from Regulatory Guide 1.97 and provide supporting justification or alternatives.

Subsequent to the issuance of the generic letter, the NRC held regional meetings in February and March 1983, to answer licensee and applicant questions and concerns regarding the NRC policy on this subject. -

At these meetings, it was noted that the NRC review would only address exceptions taken to Regulatory Guide 1.97. Furthermore, where licensees or

  • applicants explicitly state that instrument systems conform to the regulatory guide it was noted that no further staff review would be necessary. Therefore, this report only addresses exceptions to Regulatory 2

Guide 1.97. The following evaluation is an audit-of the licensee's submittals based on the review policy described in the NRC regional meetings.

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3. EVALUATION The licensee provided responses to NRC Generic Letter 82-33 on June 29,1984 (Unit 1), March 30,1984 (Unit 2), April 10,1985 and August 8, 1986 (Units 1 and 2). This evaluation is based on these submittals.

3.1 Adherence to Reaulatory Guide 1.97 The licensee stated that compliance with Regulatory Guide 1.97 is indicated on their review checklist which summarizes each variable's compliance with the Regulatory Guide 1.97 provisions. That compliance report presents justification, modifications or ongoing evaluations that are provided as resolutions for any identified deviations. Therefore, it (

is concluded that the licensee has provided an explicit commitment on conformance to Regulatory Guide 1.97. Exceptions to and deviations from the regulatory guide are noted in Section 3.3.

3.2 Tvoe A Variables Regulatory Guide 1.97 does not specifically identify Type A variables, i.e., those variables that provide information required to permit the control room operator to take specific manually controlled safety actions.

The Itcensee classifies the following instrumentation as Type A:

1. Reactor coolant system (RCS) pressure (wide range)
2. RCS hot leg temperature (wide range)
3. RCS cold leg temperature (wide range) .
4. Steam generator level (vide range) .
5. Steam generator level (narrow range) 4
6. Pressurizer. level
7. Containment pressure (nornal range)
8. Main steamline pressure
9. Refueling water storage tank level
10. Containment water level
11. Condensate storage tank level
12. Auxiliary feedwater flow l
13. Core exit temperature
14. Core subcooling monitor i The above instrumentation meets Category 1 requirements consistent with the i requirements for Type A variables, except as noted in Section 3.3.

t 3.3 Exceptions to Reaulatory' Guide 1.97 4

The licensee identified deviations and exceptions from Regulatory Guide 1.97. These are discussed in the following paragraphs.

3.3.1 Environmental Qualification Reauirement Deviation l

In References 4 and 5, the licensee has indicated that environmental

qualification is not applicable for the following Category 1 and 2

. Instrumentation. However, no justification was submitted for this deviation. In Reference 7, the licensee stated that the instrumentation l

j* listed, along with the associated instrument loop components, are located

, outside areas that constitute a harsh environment.

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Main steamline pressure Refueling water storage tank level Condensate storage tank level Plant vent stack flow Condenser steam jet air ejector radiation Plant vent effluent radiation .

Accessible area radiation Main steam effluent radiation Turbine driven auxiliary feedwater effluent radiation Heating, ventilating and air conditioning emergency damper position--control room I Pressurizer heater breaker position

Status of standby power and other energy sources important to safety.

Based on the licensee's justification that the instrumentation listed is located in a mild environment, we find this deviation acceptable.

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Deviations other than environmental qualification for these variables I are listed elsewhere in this report.

3.3.2 Neutron Flux (Intermediate Rance) j The installed neutton flux instrumentation does not completely meet the redundancy requirements of Regulatory Guide 1.97. Both intermediate range instrument loops are ultimately powered from the same DC power supply train (Train A). The power to the instrument loops is provided by separate inverters, and the outputs of these inverters are physically separated and backed up by diesel generator A. In addition, an alternate source of power, other than the inverters, is provided to both instrument loops from a Solatron voltage regulator. The licensee is installing a third channel of wide range instrumentation to resolve ambiguity between the existing

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instrumentation should one loop fail. This new instrumentation loop,

  • however, will be powered from the same power supply train (Train A) as the two existing neutron flux monitoring loops. The licensee also states that

) the existing electrical independence of the neutron monitors is consistent j with the design criteria of the reactor protection system.

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There is a very low probability of a fault occurring that would fail the complete electrical train A (ac and dc). In the event this loss should

. eccur, only the intermediate range of neutron flux instrumentation would be j lost. No single failures such as inverters, batteries, battery chargers, 4160 volt bus faults or de bus faults would disable all the neutron flux instrumentation. Based on this evaluation, we conclude that the existing

configuration of this variable is acceptable.

3.3.4 RCS Soluble Boron Concentration Regulatory Guide 1.97 recomends Category 3 instrumentation, with a j range of 0 to 6000 ppm, for this variable. The licensee takes cre:tti for the post-accident sampling system to meet this recomendation.

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The licensee takes exception to Regulatory Guide 1.97 with respect to

! post-accident sampling capability. This exception goes beyond the scope of this review and is being addressed by the NRC as part of their review of NUREG-0737 Item II.B.3.

3.3.5 RCS Cold and Hot Leo Water Temperature The maximum indication of the instrumentation for these variables is

700*f. This is 50*f less than the Regulatory Guide 1.97, Revision 2, range guidelines (50 to 750*F).

I Regulatory Guide 1.97, Revision 3. (Reference 8) recommends a range of 50 to 700*f for these variables. The instrumentation supplied by the l

itcensee meets this recommended range and is, therefore, acceptable.

3.3.6 Coolant Level in Reactor (Unit No. 2) i l The licensee does not have usable instrumentation for this variable. i As justification, the licensee states that they participated in a pilot project for a non-invasive reactor vessel level system. This unsuccessful

demonstration led the licensee into a detailed review of commercially I

available reactor vessel level systems. The results of this ongoing review 7

indicate to the itcensee that no commercially available reactor vessel level system has been accepted by the NRC for operational use.

It is our understanding that two systems are now commercially available for reading reactor vessel level in a pressurized water reactor.

One system uses heated junction thermocouples (NUREG/CR-2527, Reference 9) -

.and the other system uses differential pressure (NUREG/CR-2628 Reference 10).

The NRC is reviewing the acceptability of this variable as part of

! their review of NUREG-0737, Item II.F.2.

3.3.7 Dearees of Subcoolino The licensee has identified degrees of subcooling as a Type A I variable. As such, it should meet Category 1 requirements. The licensee states that their core subcooling monitor meets Category 2 requirements.

The NRC is reviewing the acceptability of this variable as part of i

their review of NUREG-0737, Item II.F.2.

3.3.8 Containment Sumo Water Level i

The licensee has taken exception to the range recommended by l Regulatory Guide 1.97 for the containment level instrumentation (bottom of containment to 600,000 gallon level equivalent). The licensee has instrumentation with a minimum level indication of 62.000 gallons. The licensee considers the existing range to be adequate since the minimum level indication is limited by physical installation restraints of the float type level measurement device and no operator actions are required i

below the 62,000 gallon level.

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i The reactor cavity sump level indication would provide a diverse i method of determining a water level increase in the containment. Since no l

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cperator action is required at less than the minimum indication available with the existing range, we find this to be an acceptable deviation from Regulatory Guide 1.97.

3.3.9 Containment Isolation Valve Position The licensee has not provided redundancy for all of the containment

. ' isolation valves. Some isolation valves inside and outside containment for i the same penetration are of the same train orientation and, therefore, redundant indication is not provided. The itcensee submitted the following justification for this deviation. These valves are normally closed valves and remain closed in an accident condition until remotely opened by the sperator. The power supply for these valves is for position indication as well as for power operation of the valve motor operators. The valves are {

part of a penetration which is redundant to another penetration. At least ene of these redundant piping systems must be opened during certain accident conditions. Therefore, the power for both containment isolation valves on a penetration is from the same power supply to ensure that a single power supply failure will not inhibit both penetrations from operating. Both isolation valves for the redundant penetration are supplied power from another power source.

We find the licensee's justification acceptable. Furthermore, if during an accident condition, a single train of electrical power were to fall resulting in a loss of position indication, the operator could verify that the outside containment isolation valve is closed and containment integrity maintained. Therefore, this is an acceptable deviation from Regulatory Guide 1.97.

3.3.10 Radioactivity Concentration or Radiation Level in Circulatina primary Coolant

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. The licensee uses the post-accident sample system to measure this parameter. In a letter dated February 17. 1984, the licensee states that procedures exist wF.tch relate radionuclide concentrations to core damage.

These procedures consider physical parameters such as core temperature and i

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sample locations. Alabama Power Company will implement a calculational method to assess the extent of core damage. This method will utilize the RCS post-accident sampling system in the determination of the status of fuel cladding and the magnitude of any core damage.

Based on the alternate instrumentation provided by the licensee, we ,

. conclude that the instrumentation supplied for this variable is adequate and, therefore, acceptable.

3.3.11 Containment Hydrocen Concentration

Regulatory Guide 1.97 recommends Category 1 instrumentation for this
variable with a range of 0 to 10 percent. The licensee has installed Category 3 hydrogen analyzers that do not meet the range recommended by Regulatory Guide 1.97. The licensee considers this instrumentation acceptable because the operators energize the hydrogen recombiners based on loss of coolant accident (LOCA) indications. Hydrogen concentration is not i

a LOCA indication, it is used only as the basis for verifying the hydrogen '

removal capability of the hydrogen recombiners. In the event that the i hydrogen analyzers are unavailable to provide containment hydrogen '

! concentration, sufficient time is available to determine the containment hydrogen concentration utt112ing the containment air post-accident sampling system (CAPASS).

The NRC has reviewed the acceptability of this variable as part of their review of NUREG-0737. Item II.F.1.6.

3.3.12 Residual Heat Removal (RHR) Heat Exchanaer Outlet Temperature Regulatory Guide 1.97, Revision 2, recommends a range for this variable of 32 to 350*F. Revision 3 changed the recommended range to 40 to 350*F. The licensee has supplied a range of 50 to 400*F. The instrumentation supplied has a range where the lower limit of the span does not conform to either revision of the regulatory guide. The licensee 10

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, states that thn existing range of this instrumentation envelops the RHR 3

system design parameters and no need exists for temperature indication below 50*f.

i j Based on the justification provided by the licensee, we conclude that the instrumentation supplied for this variable is adequate to monitor this variable during all accident and post-accident conditions, and is

, therefore, acceptable.

3.3.13 Accumulator Tank Level and Pressure [

i The licensee has Category 3 accumulator tank level instrumentation that does not meet the recommended range. The justification submitted by l the licensee for this deviation is that the accumulator tank level at the l Farley Nuclear Plant was designed solely to verify compliance with the

! technical specification volume provisions. In the event of RCS

depressurization, accumulator tank discharge is verified by monitoring accumulator tank pressure, which mects the Category 2 requirements. -

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j The accumulators are passive and discharge for RCS breaks. The level i ar,d pressure measurement channels are not required to protect the integrity l

of the RCS boundary, to shutdown the reactor, to maintain it in a safe j shutdown condition or to prevent or mitigate the consequences of an j accident which could result in potential exposures. We find the qualified

! pressure instrumentation supplied for this variable adequate to determine l that the accumulators have discharged. Therefore, the existing l instrumentation is acceptable to monitor this variable.

3.3.14 Refuelina Water Storace Tank Level Regulatory Guide 1.97 recommends a range from the top to the bottom of l the tank. The licensee does not meet this range and states that the

i. maximum level indication of the existing instrumentation is one foot below j the top of the tank. This level indication reads from 0 to 40 feet and '

j cnvelopes the technical specification volume requirement, which the licensee states is sufficient to mitigate any design basis event. This l

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range is adequate to provide the operator with information for normal operations and to perform switchover from emergency core cooling system injection to recirculation.

Based on the licensee's justification, we conclude that the existing instrumentation for this variable (that reads 98 percent of the recommended

. range) is adequate to monitor this variable during all accident and post-accident conditions.

3.3.15 Pressurizer Level Regulatory Guide 1.97 reconmends a range from the bottom to the top for this variable. The instrumentation provided by the licensee does not read this full range. The licensee states that the volume measured represents approximately 89 percent of the pressurizer and is sufficient for the operator to take the required manual actions and to ensure the proper operation of the pressurizer.

The portion of the pressurizer level that is not indicated (approximately 11 percent) is the upper and lower hemispherical head region, where the volume to level ratio is not linear. We find this deviation minor and acceptable. The existing range is adequate to monitor this variable during all accident and post-accident conditions.

3.3.16 Pressurizer Heater Status i

Regulatory Guide 1.97 reconnends electric current instrumentation to determine the operating status of the pressurizer heaters. The licensee does not intend to provide specific instrumentation to read this current, i The licensee states, in Reference 4 and 5, that the pressurizer heater status can be adequately determined by the use of pressurizer heater circuit breaker position and pressurizer pressure, furthermore, the licensee states that the emergency response procedures no not uttitre pressurizer heater current for accident mitigation.

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In Reference 7, the licensee lists several additional means of i

determining pressurizer heater operation. In addition to these, the licensee states that heaters current can be monitored with the diesel generator megawatt and current indicators when the diesel generator is supplying power for the heaters and with the 4.16 KV bus incoming current indicator when being supplied from offsite power.

, Based on the available current monitoring instrumentation and the diverse means of determining heater operation, we conclude that this is an j acceptable deviation from Regulatory Guide 1.97.

3.3.17 Quench Tank Level The range of the existing instrumentation for this variable does not 3

meet the range recommended by Regulatory Guide 1.97 (top to bottom). The j licensee's justification for this deviation is that only 5 percent of the total tank volume is not measured and the existing range is sufficient to provide the operator with the necessary information for accident monitoring.

We find the existing level range adequate to monitor the operation of this tark. Therefore, this is an acceptable deviation from Regulatory Guide 1.97.

3.3.18 1_ team Generator Level Regulatory Guide 1.97 recommends a range from the tube sheet to the separators for this variable. The licensee has instrumentation that reads from 12 in, above the tube sheet to the separators. The licensee states that the volume of the steam generator not measured is less than 2 percent of the volume recommended by the regulatory guide.

The steam generator is, in effect, empty at 12 in, above the tube

. sheet; therefore, this deviation is minor with respect to the overall range j and system accuracy. The existing range is adequate to monitor this i variable during all accident and post-accident conditions.

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1 3.3.19 Steam Generator Pressure i

1 The licensee has instrumentation for this variable with a range of i

! O to 1200 psig. Regulatory Guide 1.97 recommends a range from atmospheric pressure to 20 percent above the lowest safety valve setting. Thus, the range should be to 1290 psig, as the licensee identifies the range as being .

90 psig less than recommended. The licensee justifies the range deviation i by stating that the highest actuation setpoint of the main steam safety I valves is 1129 psig. Allowing for 3 percent accumulation the maximum

! credible steamline pressure is 1163 psig which is within the indicated range of the existing instrumentation.

4 Based on the licensee's statement that the maximum credible steamline l pressure would be 1163 psig, this instrumentation would remain on scale l during any accident or post-accident conditions. Therefore, we find this

) deviation acceptable.

3.3.20 Volume Control Tank Level The itcensee takes exception to the range recommended by Regulatory f

l Guide 1.97 for this variable (top to bottom). The transmitters measure the full range between the instrument connections, however, these connections l are not at the top and bottom of the tank. The justification submittad by the licensee is that for operational purposes, level indication at either 1

end of the scale is considered full or empty. Also, the existing range of l

the volume control tank level envelops all automatic action of the level l control system.

We find that the existing level indication is adequate to monitor the

operation of this tank. Therefore, this is an acceptable deviation from '

Regulatory Guide 1.97.

! 3.3.21 Hinh Level Radioactive Linuld Tank Level

! Regulatory Guide 1.97 recommends a range of top to bottom for this variable. The transmitters for these tanks measure the full range between i 14 i

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the instrument connections; however, these connections are not at the top I}

and bottom of the tank. The licensee's justification for this deviation is that at least 90 percent of the tank volume is measured and the range is sufficient to provide the operator with the necessary information for accident monitoring.

i We find that the existing range is adequate to monitor the operation j, ,cf this tank during all accident and post-accident conditions. Therefore, this is an acceptable deviation from Regulatory Guide 1.97. i i

3.3.?? Radioactive Gas HolduD Tank Pressure 1

i l The licensee takes exception to the range recommended by Regulatory Guide 1.97 for this variable (0 to 150 percent design pressure). The licensee has instrumentation for this variable that reads from 0 to l 100 percent of the design pressure of the tank (150 psig). The licensee l

states that the existing range is acceptable because it covers up to the i design pressure of the tanks and because relief valves are installed on I each tank to prevent the tank pressure from exceeding the design value of I 150 psig.

l Based on the justification provided by the licensee, we conclude that the instrumentation provided for this variable is adequate to monitor the i cperation of this tank and is, therefore, acceptable.

3.3.23 Radiation Exoosure Rate a

i j In References 4.and 5, the licensee identified two deviations for this variable. First, of the plant areas which are accessible post-accident, l only the control room has a permanently installed radiation monitor.

, Second, the range of the radiation level indication of the control room radiation monitor is 10'* to 10 R/hr. The range specified by Regulatory Guide 1.97 for this variable is 10'I to 104 R/hr.

I i l In Reference 7, the licensee identified permanently installed monitors i in areas required for post-accident access. The ranges of these i

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instruments are less than recommended by the regulatory guide, however the licensee states that the ranges are adequate since additional shielding has been installed in these areas and portable monitoring instruments would be used to assess radiation levels before entry into these areas.

The licensee has shown an analysis of radiation levels expected for the monitor locations. The existing radiation exposure rate monitors have -

' ranges that encompasses the expected radiation levels in their location.

Based on this, and the fact that personnel would not be permitted into the area without portable monitoring if the upper limit of the range is exceeded, we find the instrumentation provided for this variable acceptable.

3.3.24 Plant and Environs Radiation (Portable Instrumentation)

Regulatory Guide 1.97 recommends a range of 10-3to 10" rads /hr, for beta radiation and low energy photons. The licensee states that the maximum indication of the existing portable instrumentation is below the recommended maximum level. The Itcensee's justification for this deviation .

is that their portable instrumentation has sufficient range to monitor the radiation levels in areas of the plant where post-accident access is necessary by plant personnel.

This instrumentation is portable and would not be used to assess levels of radiation greater than the range provided by the licensee.

Therefore, this is an acceptable deviation from Regulatory Guide 1.97.

3.3.25 PlantandEnvironsRadioactivity(Portableinstrumentation)

The licensee does not have a portable multichannel gamma-ray spectrometer, as recommended by Regulatory Guide 1.97, Revision 2, for this variable. Regulatory Guide 1.97, Revision 3. states that portable instrumentation should be provided for isotopic analysis of plant and environs radioactivity. The licensee has also not provided portable ,

instrumentation for isotopic analysis. However, the licensee does have two non-portable multichannel analyzers (MCA) located in the counting room of 16

r the plant. The MCAs are equipped with a germanium-lithium detector to provide isotopic analysis of the plant and environ samples. The MCAs have the capability to analyze samples in less than 15 minutes from the time the sample is delivered to the MCAs. The MCAs located in the plant are used during normal plant operations, are accessible post-accident, and are instruments familiar to plant personnel.

. The licensee states that a portable multichannel gamma-ray spectrometer would not enhance the capability to perform isotopic analysis. A portable device can only provide " scoping" of the radionuclide content and cannot provide a quantitative measurement. The existing non-portable MCAs at the Farley Nuclear Plant would provide a quantitative r.easurement of the radionuclide content.

The two existing multichannel analyzers are sufficient to provide for isotopic analysis and an adequate and timely assessment of radioactive releases at this station. Therefore, this is an acceptable deviation from Regulatory Guide 1.97. .

3.3.26 Wind Speed Regulatory Guide 1.97, Revision 2, recommends a range of 0 to 30 meters /second (67 mph) for this variable. The licensee has j instrumentation with a range of 0 to 22 meters /second (50 mph). The licensee justifies this deviation by stating that their existing wind speed

} instrumentation has historically provided reliable indications that are l representative of meteorological conditions in the plant vicinity.

Regulatory Guide 1.97, Revision 3, reconnends instrumentation with a range of 0 to 22 meters /second (50 mph) for this variable. Since the f existing instrumentation meets the Regulatory Guide 1.97, Revision 3

!* requirement, this deviation is acceptable.

3.3.27 Accident Samotina (Primary Coolant. Containment Air and Sumoi The minimum quantifiable concentrations of boron, chlorides, dissolved hydrogen, total gas and oxy;en do not meet Regulatory Guide 1.97 range

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guidelines. The licensee states that analysis below the minimums '

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identified would serve no useful purpose for accident analysis, mitigation or recovery.

The minimum quantifiable concentrations of oxygen and hydrogen in the containment air do not satisfy Regulatory Guide 1.97 range guidelines. The licensee': justification for this deviation is that the minimum

,quantifiable concentrations represent the minimum detectable concentrations. In addition, the licensee states that analysis below the identified minimums would serve no useful purpose for accident analysis, mitigation or recovery.

The licensee takes exception to Regulatory Guide 1.97 with respect to post-accident sampling capability. These exceptions go beyond the scope of this review and are being addressed by the NRC as part of their review of NUREG-0737, Item II.B.3.

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4. CONCLUSIONS Based on our review, we find that the licensee conforms to, or is justified in deviating from Regulatory Guide 1.97.

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5. REFERENCES
1. NRC letter, D. G. Eisenhut to All Licensees of Operating Reactors, Applicants for Operating Licenses, and Holders of Construction Permits, " Supplement No. 1 to NUREG-0737--Requirements for Emergency Response Capability (Generic Letter No. 82-33)," December 17, 1982.
2. Instrumentation for Licht-Water-Cooled huclear Power Plants to Assess Plant and Environs Conditions Durina and Followina an Accident,
3. Clarification of TMI Action Plan Reautrements. Reautrements for Emeraency Response Capability, NUREG-0737, Supplement No. 1, NRC, Office of Nuclear Reactor Regulation, January 1983.
4. Alabana Power Company letter, R. P. Mcdonald to Director, Nuclear Reactor Regulation, NRC, " Regulatory Guide 1.97 Compliance," June 29, 1984.
5. Alabana Power Company letter, F. L. Clayton, Jr. to Director, Nuclear Reactor Regulation, NRC, " Regulatory Guide 1.97 Compliance,"

March 30, 1984.

6. Alabana Power Company letter, R. P. Mcdonald to Director, Nuclear Reactor Regulation, NRC, " Regulatory Guide 1.97 Compliance," April 10, ,

1985.

7. Alabama Power Company letter, R. P. Mcdonald to Director, Nuclear Reactor Regulation NRC, " Regulatory Guide 1.97 Compliance",

August 8, 1986.

8. Instrumsntation for Licht-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions Durina and followina an Accident, Regulatory Guide 1.97, Revision 3 NRC, Office of Nuclear Regulatory Research, May 1983.
9. 'nadeauate Core Coolina Instrumentation Usina Heated Junction
hermocouples for Reactor Vessel level Measurement, NUREG/CR-2627, ORNL/IM-8248, March 1982.
10. Inadeauste Core Coolina Instrumentation Usina Differential Pressure for Reactor Vessel Level Measurment, NUREG/CR-2628, ORNL/TM-8269, March 1982.

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Washington, DC 20555

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This EGSG Idaho, Inc. report reviews the submittals for the Joseph fl. Farley Nuclear Plant, Unit Nos.1 & 2, and identifies areas of non-conformance to Regulatory Guide 1.97. Any exceptions to these guidelines are evaluated.

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